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UNITED STATES
NUCLEAR REGULATORY COMMISSION
REGION I
475 ALLENDALE ROAD
KING OF PRUSSIA, PENNSYLVANIA 19406·1415
April 28, 2010
Mr. Charles G. Pardee
Senior Vice President, Exelon Generation Company LLC
President and Chief Nuclear Officer, Exelon Nuclear
4300 Winfield Rd.
Warrenville, IL 60555
SUBJECT:
PEACH BOTTOM ATOMIC POWER STATION, UNITS 2 AND 3 - NRC
EVALUATION OF CHANGES, TESTS, AND EXPERIMENTS AND
PERMANENT MODIFICATIONS TEAM INSPECTION REPORT
05000277/2010006 AND 05000278/2010006
Dear Mr. Pardee:
On March 19, 2010, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection
at the Peach Bottom Atomic Power Station. The enclosed inspection report documents the
inspection results, which were discussed on March 19, 2010, with Mr. W. Maguire, and other
members of your staff.
The inspection examined activities conducted under your license as they relate to safety and
compliance with the Commission's rules and regulations and with the conditions of your license.
In conducting the inspection, the team reviewed selected procedures, calculations and records,
observed activities, and interviewed station personnel.
Based on the results of this inspection, no findings of significance were identified.
In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter, its
enclosure, and your response (if any) will be available electronically for public inspection in the
NRC Public Document Room or from the Publicly Available Records (PARS) component of the
NRC's document system (ADAMS). ADAMS is accessible from the NRC Web site at
http://www.nrc.gov/reading-rm/adams.html(the Public Electronic Reading Room).
Sincerely,
c:Yc~~ -f 6o&JOo~
Chief~! ~
Lawrence T. Doerflein,
Engineering Branch 2
Division of Reactor Safety
Docket Nos.: 50-277; 50-278
License Nos.: DPR-44, DPR-56
Enclosure:
Inspection Report No. 05000277/2010006 and 05000278/2010006
w/Attachment: Supplemental Information
cc w/encl: Distribution via ListServ
April 28, 2010
Mr. Charles G. Pardee
Senior Vice President, Exelon Generation Company LLC
President and Chief Nuclear Officer, Exelon Nuclear
4300 Winfield Rd.
Warrenville, IL 60555
SUBJECT:
PEACH BOTTOM ATOMIC POWER STATION, UNITS 2 AND 3 - NRC
EVALUATION OF CHANGES, TESTS, AND EXPERIMENTS AND
PERMANENT MODIFICATIONS TEAM INSPECTION REPORT
05000277/2010006 AND 05000278/2010006
Dear Mr. Pardee:
On March 19, 2010, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection
at the Peach Bottom Atomic Power Station. The enclosed inspection report documents the
inspection results, which were discussed on March 19,2010, with Mr. W. Maguire, and other
members of your staff.
The inspection examined activities conducted under your license as they relate to safety and
compliance with the Commission's rules and regulations and with the conditions of your license.
In conducting the inspection, the team reviewed selected procedures, calculations and records,
observed activities, and interviewed station personnel.
Based on the results of this inspection, no findings of significance were identified.
In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter, its
enclosure, and your response (if any) will be available electronically for public inspection in the
NRC Public Document Room or from the Publicly Available Records (PARS) component of the
NRC's document system (ADAMS). ADAMS is accessible from the NRC Web site at
http://www.nrc.gov/reading-rm/adams.html(the Public Electronic Reading Room).
Sincerely,
IRA!
Lawrence T. Doerflein, Chief
Engineering Branch 2
Division of Reactor Safety
Docket Nos.: 50-277,50-278
License Nos.: DPR-44, DPR-56
Enclosure:
Inspection Report No. 05000277/2010006 and 05000278/2010006
w/Attachment: Supplemental Information
cc w/encl: Distribution via ListServ
ADAMS ACCESSION NO. ML 101180465
SUNSI Review Complete: LTO
(Reviewer'S Initials)
DOCUMENT NAME: G:\DRS\Engineering Branch 3\Lilliendahl\PB Mods Report 2010006 rev1.doc
After declaring this document ~An Official Aaencv Record- it will be released to the Public. To receive a copy of this document, Indicate In the box: ~C";;; C()py__ 'Nithout attachmenVenclosure "E";;; COpy INith attachment/enclosure "N"
OFFICE
RIIDRS
NAME
DATE
JLilliendahl1
04/14/10
I
RI/DRP
PKrohnl
04/15/10
I
RI/DRS
LDoerfJeinl
04/28/10
OFFICIAL RECORD COpy
I
I
=No copy
I
C. Pardee
2
Distribution w/encl: (via E-mail)
S. Collins, RA
(R10RAMAIL Resource)
M. Dapas, ORA (R10RAMAIL Resource)
D. Lew, DRP
(R1DRPMAIL Resource)
J. Clifford, DRP (R1DRPAMAIL Resource)
D. Roberts, DRS (R1DRSMail Resource)
P. Wilson, DRS (R1DRSMail Resource)
P. Krohn, DRP
A. Rosebrook, DRP
E. Torres, DRP
J. Bream, DRP
F. Bower, DRP, SRI
A. Ziedonis, DRP, RI
S. Schmitt, DRP, OA
L. Trocine, RI OEDO
D. Bearde, DRS
RidsNrrPMPeachBottom Resource
[email protected]
L. Doerflein, DRS
J. Ulliendahl, DRS
u.s. NUCLEAR REGULATORY COMMISSION
REGION I
Docket No.:
50-277 and 50-278
License No.:
DPR-44 and DPR-56
Report No.:
05000277/2010006 and 05000278/2010006
Licensee:
Exelon Generation Company, LLC (Exelon)
Facility:
Peach Bottom Atomic Power Station, Units 2 and 3
Location:
Delta, Pennsylvania
Inspection Period:
March 1, 2010 through March 19, 2010
Inspectors:
J. Lilliendahl, Reactor Inspector, Division of Reactor Safety (DRS),
Team Leader
K. Mangan, Senior Reactor Inspector, DRS
M. Orr, Reactor Inspector, DRS
S. Rich, Reactor Engineer, NSPDP (in-training)
G. Figueroa, Reactor Systems Engineer, Office of Nuclear Regulatory
Research (RES) (observer)
Approved By:
Lawrence T. Doerflein, Chief
Engineering Branch 2
Division of Reactor Safety
Enclosure
SUMMARY OF FINDINGS
IR 05000277/2010006 and 05000278/2010006; 3/1/2010 - 3/19/2010; Peach Bottom Atomic
Power Station, Units 2 and 3; Engineering Specialist Plant Modifications Inspection.
This report covers a two week on-site inspection period of the evaluations of changes, tests, or
experiments and permanent plant modifications. The inspection was conducted by three region
based engineering inspectors, one inspector in training, and one observer. No findings of
significance were identified. The NRC's program for overseeing the safe operation of
commercial nuclear power reactors is described in NUREG-1649, "Reactor Oversight Process,"
Revision 4, dated December 2006.
A.
NRC-Identified and Self-Revealing Findings
No findings of significance were identified.
B.
Licensee-Identified Violations
None.
ii
Enclosure
REPORT DETAILS
1.
REACTOR SAFETY
Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity
1R17 Evaluations of Changes, Tests, or Experiments and Permanent Plant Modifications
(IP71111.17)
.1
a.
Evaluations of Changes, Tests, or Experiments (22 samples)
Inspection Scope
The team reviewed six safety evaluations to determine whether the changes to the
facility or procedures, as described in the Updated Final Safety Analysis Report
(UFSAR), had been reviewed and documented in accordance with 10 CFR 50.59
requirements. In addition, the team evaluated whether Exelon had been required to
obtain NRC approval prior to implementing the change. The team interviewed plant
staff and reviewed supporting information including calculations, analyses, design
change documentation, procedures, the UFSAR, the Technical Specifications (TSs),
and plant drawings, to assess the adequacy of the safety evaluations. The team
compared the safety evaluations and supporting documents to the guidance and
methods provided in Nuclear Energy Institute (NEI) 96-07, "Guidelines for 10 CFR 50.59
Evaluations," as endorsed by NRC Regulatory Guide 1.187, "Guidance for
Implementation of 10 CFR 50.59, Changes, Tests, and Experiments," to determine the
adequacy of the safety evaluations.
The team also reviewed a sample of sixteen 10 CFR 50.59 screenings for which Exelon
had concluded that no safety evaluation was required. These reviews were performed
to assess whether Exelon's threshold for performing safety evaluations was consistent
with 10 CFR 50.59. The sample included design changes, calculations, and procedure
changes.
The team reviewed the safety evaluations that Exelon had performed and approved
during the time period covered by this inspection (Le., since the last modifications
inspection). The screenings and applicability determinations were selected based on
the safety significance, risk significance, and complexity of the change to the facility.
In addition, the team compared Exelon's administrative procedures used to control the
screening, preparation, review, and approval of safety evaluations to the guidance in
NEI 96-07 to determine whether those procedures adequately implemented the
requirements of 10 CFR 50.59. The reviewed safety evaluations and screenings are
listed in the attachment.
b.
Findings
No findings of significance were identified.
Enclosure
2
.2
Permanent Plant Modifications (10 samples)
.2.1
High Pressure Service Water (HPSW) Orifice Changes
a.
Inspection Scope
The team reviewed a modification (Engineering Change Request (ECR) 06-00023) to
the HPSW system outlet piping from the Unit 3 '0' residual heat removal (RHR) heat
exchanger. The modification installed a multi-hole orifice plate and removed the plugs
from another orifice plate in the same line. This change was necessary following the
replacement of internal components of the Unit 3 '0' RHR heat exchanger outlet valve,
because the newer assembly had a greater resistance coefficient and resultant pressure
drop. The team conducted the review to verify that the design bases, licensing bases,
and performance capability of the HPSW system had not been degraded by the
modification.
The team discussed the modification and design basis with design and system
engineers to assess the adequacy of the modification and to verify that the HPSW
system had been fully restored to service. Post modification test results were reviewed
to ensure the HPSW pressure and flow rate met the acceptance criteria. The 10 CFR
50.59 screening determination associated with this modification was also reviewed as
described in section 1R17.1 of this report. Additional documents reviewed are listed in
the attachment.
b.
Findings
No findings of significance were identified .
.2.2
a.
'E1' Emergency Diesel Generator (EDG) Speed Switch Replacement
Inspection Scope
The team reviewed a modification (ECR 07-00284) which replaced the 'E l' EDG engine
speed switch. Exelon implemented this modification following erratic behavior of the
speed switch during routine surveillance testing. Exelon had previously upgraded the
speed switch to a newer model, and did not have any available new spares. The
modification re-installed the original model speed switch. The original model speed
switch had a successful history of use since plant construction and was determined to
meet the criteria of an equivalent change. The team conducted the review to ensure
that the design bases, licensing bases, and performance capability of the EDG had not
been adversely affected by the modification.
The team reviewed the high and low speed switch setpoints to verify there were no
adverse impacts to operating margins and that the speed switch functioned in
accordance with the design basis. The team reviewed post maintenance test data to
confirm that the replacement speed switch met the acceptance criteria and the EDG
Enclosure
3
was operable. Additionally, the team reviewed the associated work order packages and
conducted interviews with design and system engineers regarding the design,
installation, and testing of the switches to verify that the modification was adequate.
The team walked down the 'E1' EDG to assess the system configuration. The
documents reviewed are listed in the attachment.
b.
Findings
No findings of significance were identified .
.2.3
a.
'E1' EDG Tachometer Drive Coupling Shaft/Pin Replacement
Inspection Scope
The team reviewed a modification (ECR 09-00360) associated with the replacement of
the 'E1' EDG tachometer drive coupling shaft and pins. These components support the
EDG's automatic start design function. Exelon implemented this modification following
the determination that the tachometer drive coupling shaft and pins had been damaged
during engine reassembly. The replacement drive assembly did not include the required
shaft and pins, which transfer camshaft rotation through the tachometer drive to the
speed switch. This modification consisted of the fabrication of a new shaft and pins
identical to the existing components. The team evaluated the change to confirm that the
design bases, licensing bases, and performance capability of the EDG had not been
inadvertently degraded by the modification.
The team reviewed the plan to fabricate the replacement parts to evaluate the suitability
of the replacement material. The team interviewed design engineers with regards to the
technical evaluation of material used, material specifications, risk assessment,
.fabrication and acceptable machining tolerances, single point vulnerability review, and
acceptance testing. The team reviewed performance test data obtained after the
installation of the tachometer drive coupling and the on-site manufactured shaft and pins
to verify proper performance of the 'E1' EDG speed switch. The team performed a
walkdown of 'E1' EDG to assess the system configuration. The documents reviewed
are listed in the attachment.
b.
Findings
No findings of significance were identified .
.2.4
a.
Core Spray Pump Technical Specification and In-service Test Acceptance Criteria
Inspection Scope
The team reviewed the design change (ECR 08-00074) that revised the acceptance
criteria for the core spray pump surveillance tests. Exelon revised calculation ME-0073,
Core Spray Pump In-Service Testing Acceptance Criteria, as well as the TS surveillance
Enclosure
4
bases documents. Exelon implemented this design change to ensure pump testing
results could be used for both in-service testing (1ST) requirements for differential
pressure and TS surveillance requirements which used discharge pressure as an
acceptance criterion. The team evaluated the change to ensure that the design bases
and licensing bases had not been adversely affected, and performance capability of the
core spray system would be properly evaluated against design and licensing
requirements.
The team reviewed calculation ME-0073 to determine if the pressure acceptance
criterion for each pump had been correctly calculated. Additionally, the team verified the
appropriateness of calculation assumptions. The team also reviewed the core spray
pump surveillance procedures to confirm that the design change had been incorporated
into the procedures and the test configuration ensured design requirements were
satisfied. The team assessed past surveillance test data to ensure that any negative
trends would be captured by the new criteria. Finally, the team interviewed design and
1ST engineers to discuss the design criteria associated with the change. Documents
reviewed are listed in the attachment.
b.
Findings
No findings of significance were identified .
.2.5
a.
Modification of MO-3-23-016 Torque Switch
Inspection Scope
The team reviewed a modification (ECR 07-00342) that changed the closing logic for the
high pressure coolant injection (HPCI) turbine steam line isolation valve, MO-3-23-016.
The valve also has the safety function to close for primary containment isolation. The
modification replaced the 2-rotor switch with a 4-rotor switch and bypassed the torque
switch stop signal until the valve closed indication was also actuated. Exelon
implemented this modification in order to allow the use of the full capability of the valve
motor to close the valve rather than be limited by the torque switch setting. The team
conducted the review to ensure that the design bases, licensing bases, and
performance capabilities of the primary containment isolation system and HPCI system
had not been adversely affected by the modification.
The team assessed selected design inputs and attributes to ensure that they were
consistent with the design and licensing bases. The team verified that the safety-related
component qualification for the new switch was adequate. Additionally, the team
confirmed that the new design did not introduce any new failure modes for the valve
which could impact the HPCI system design basis accident response. The team also
reviewed the post modification test plan and results to ensure appropriate acceptance
criteria had been met which demonstrated the adequacy of the new design. Finally, the
team interviewed the motor operated valve engineer to discuss the implementation of
the modification. The 10 CFR 50.59 screening determination associated with this
Enclosure
5
modification was also reviewed as described in section 1R17.1 of this report.
Documents reviewed are listed in the attachment.
b.
Findings
No findings of significance were identified .
.2.6
a.
Implementation of New Technical Specifications for Control Room Envelope
Inspection Scope
The team reviewed a modification (ECR 08-00104) to the control room envelope (CRE)
habitability program that changed the way operability of the CRE was evaluated. With
the change Exelon established a maximum breech size in the envelope as the
requirement to ensure its operability. Exelon performed a calculation to determine the
size of the maximum opening, based on the capabilities of the main control room
emergency ventilation (MCREV) system, as well as the hole size that correlated to the
current leakage out of the CRE. Exelon implemented this modification in order to more
easily determine the impact of maintenance work on CRE operability. The team
conducted the review to ensure that the design bases, licensing bases, and
performance capabilities of the CRE and MCREV system had not been adversely
affected by the modification.
The team reviewed calculation PM-1081, Calculation of ~aximum CRE Breach, to
determine that the maximum opening size had been correctly calculated and that
assumptions used in the calculation were appropriate. The team also reviewed
completed copies of procedure GP-30, Control Room Envelope Boundary Integrity, to
verify that the opening size limits were being maintained. The team interviewed the
design engineer to discuss the design criteria associated with the change, and walked
down the CRE with the program engineer to ensure that conditions in the field aligned
with the calculation assumptions. Documents reviewed are listed in the attachment.
b.
Findings
No findings of significance were identified .
.2.7
a.
Unit 3 Cycle 16 Core Reload Analysis
Inspection Scope
The team reviewed the analysis performed by Exelon for the design of the Unit 3 cycle
16 core reload (ECR 07-00158). The analysis was performed to ensure that various
core parameters and thermal limits would not be exceeded throughout the life of the
core. The team conducted the review to ensure that the design bases and licensing
bases for the fuel had not been adversely affected by the modification. The team
reviewed the analysis to determine what codes were used in the core design and
Enclosure
6
compared those codes to the codes approved by the NRC as listed in the UFSAR. The
team also reviewed any changes made by the fuel vendor to the approved codes to
ensure that changes made to the approved methodologies resulted in conservative
results. Additionally, the team reviewed the results from start-up physics testing to verify
that appropriate acceptance criteria had been applied and the testing results were within
the acceptance criteria. Documents reviewed are listed in the attachment.
b.
Findings
No findings of significance were identified .
.2.8
a.
Modification to Provide Continuous Prelube for 'E1' EDG
Inspection Scope
The team reviewed a modification (ECR 05-00255) which provided continuous prelube
of the 'E1' EDG lower crankshaft line. Exelon implemented the modification to keep the
system piping, filters, strainers, and heat exchangers filled so that lube oil will be quickly
provided to the bearings in the event of an emergency start. The modification included
replacing the existing lube oil heater with a higher kilowatt rated heater, and adding a
thermostatically controlled valve that allows oil to flow to the engine lower crankshaft line
only after it has reached a sufficient temperature. The team conducted the review to
ensure that the design bases, licensing bases, and performance capability of the EDG
had not been adversely affected by the modification.
The team assessed selected design inputs and attributes to ensure that they were
consistent with the design and licensing bases. The team interviewed the responsible
design and system engineers to understand the implementation of the modification and
subsequent system performance. The team reviewed the electrical loading calculations
and breaker coordination calculations to assess the adequacy of the design with respect
to the additional heater load. Additionally, the team evaluated if the design introduced
any new failure modes for the lube oil system which could impact EDG design basis
accident response. The team also reviewed the post modification test plan and results
to ensure appropriate acceptance criteria had been met and the tests demonstrated the
adequacy of the new design. The 10 CFR 50.59 screening determination associated
with this modification was also reviewed as described in section 1R17.1 of this report.
Documents reviewed are listed in the attachment.
b.
Findings
No findings of significance were identified.
Enclosure
7
.2.9
a.
Modification to Replace Electro-Mechanical Trip Devices with Solid State Trip Devices
Inspection Scope
The team reviewed a modification (ECR 07-00001) which replaced the electromechanical trip devices in safety related 480V breakers with solid state trip devices.
Exelon implemented the modification because the electro-mechanical devices were
obsolete. The modification included replacing the trip devices, setting the new trip
device setpoints, and updating the breaker coordination calculation. The breaker
coordination calculation was updated to verify that the new device trip characteristics
would provide adequate coordination. The team conducted the review to ensure that
the design bases, licensing bases, and performance capability of the safety related
480V system had not been adversely affected by the modification.
The team assessed selected design inputs and attributes to ensure that they were
consistent with the design and licensing bases. The team interviewed the responsible
design engineer to understand the implementation of the modification and subsequent
system performance. The team reviewed the original and revised versions of PE-0193,
Coordination Study for 480VAC Load Centers and Motor Control Centers (MCCs), to
assess the adequacy of the design with respect to the new trip device characteristics
and the new trip settings. The team also reviewed the post modification test plan and
results to ensure that the actual trip settings were in accordance with the design trip
settings. The associated equivalency change evaluation was also reviewed.
Documents reviewed are listed in the attachment.
b.
Findings
No findings of significance were identified .
.2.10
a.
Modification to Retire 'E3' EDG Auxiliary Pumps and Associated Breaker Change
Inspection Scope
The team reviewed a modification (ECR 08-00017) that retired the 'E3' EDG jacket
coolant, air coolant, and lube oil motor driven backup auxiliary pumps. Exelon
implemented the modification because the aforementioned pumps required excessive
maintenance, introduced leak paths, increased the failure modes, and did not
appreciably increase the reliability of the EDGs. Exelon had previously completed this
modification for the other EDGs. The modification included isolating and removing the
pumps (blind flanges, etc.), removing the pump power and controls, and replacing the
EDG MCC instantaneous magnetic molded case circuit breakers (MCCBs) with thermal
magnetic MCCBs. The team conducted the review to ensure that the design bases,
licensing bases, and performance capability of the EDG had not been adversely
affected by the modification.
.
Enclosure
8
The team assessed selected design inputs and attributes to ensure that they were
consistent with the design and licensing bases. The team interviewed the responsible
design and system engineers to understand the implementation of the modification and
subsequent system performance. The team reviewed the breaker coordination
calculations to assess the adequacy of the design with respect to the change in circuit
breaker design. Additionally, the team confirmed that the design did not introduce any
new failure modes for the lube oil system, jacket water system, or air cooling system
which could impact EDG design basis accident response. The team also reviewed the
post modification test plan and results to ensure appropriate acceptance criteria had
been met and the tests demonstrated the adequacy of the new design. Finally, the
team walked down the 'E3' EDG to verify that the retired equipment was fully isolated
from the EDG. The 10 CFR 50.59 screening determination associated with this
modification was also reviewed as described in section 1R 17.1 of this report.
Documents reviewed are listed in the attachment.
b.
Findings
No findings of significance were identified.
4.
OTHER ACTIVITIES
40A2 Identification and Resolution of Problems (IP 71152)
a.
Inspection Scope
The team reviewed a sample of Issue Reports (IRs) associated with 10 CFR 50.59 and
plant modification issues to determine whether Exelon was appropriately identifying,
characterizing, and correcting problems associated with these areas, and whether the
planned or completed corrective actions were appropriate. In addition, the team
reviewed IRs written on issues identified during the inspection to verify adequate
problem identification and incorporation of the problem into the corrective action system.
The IRs reviewed are listed in the attachment.
b.
Findings
No findings of significance were identified.
40A6 Meetings. including Exit
The team presented the inspection results to Mr. W. Maguire, Site Vice President, and
other members of Exelon's staff at an exit meeting on March 19,2010. The team
returned the proprietary information reviewed during the inspection to the licensee and
verified that this report does not contain proprietary information.
Enclosure
A-1
ATTACHMENT
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Exelon' Personnel
W. Maguire, Site Vice President
R. Smith, Licensing
J. Chizever, Manager, Mechanical Design Engineering
K. Cutler, Design Engineer
LIST OF ITEMS OPENED, CLOSED AND DISCUSSED
None.
LIST OF DOCUMENTS REVIEWED
10 CFR 50.59 Evaluations
PB-2007-001-E,
PB-2007-002-E,
PB-2007 -003-E,
PB-2008-001-E,
PB-2009-001-E,
PB-2009-002-E,
Improved Control Rod Shutdown Sequence (Soft Shutdown), Rev. 0
GP-3 Revision to Defeat the Mode Switch to Shutdown Scram, Rev. 0
Continuous Venting of Recirculation Seals, Rev. 0
Unit 2 Drywell Equipment Drain Sump Temporary Logic Reconfiguration, Rev. 0
Install Higher Value Sensitivity Resistor on Control Circuit LS-2-20-360, Rev. 0
Application of TRACG04P for OPRM Setpoint Determination, Rev. 0
10 CFR 50.59 Screened-Out Evaluations
PB-2005-065-S, PBAPS EDG Keep Warm Modifications, Rev. 5
PB-2006-055-S, Abandonment of EDG Motor-Driven Aux Lube Oil, Jacket Coolant, and Air
Coolant Pumps, Rev. 0
PB-2007-007-S, ESW Pump Discharge Cross-Tie Operation, Rev. 0
PB-2007-008-S, Low Flow Thru HPSW Side of 3D RHR HX During RT-O-032-300-3, Rev. 0
PB-2007-028-S, Revise RCIC TDH Requirements in UFSAR and DBD, Rev. 0
PB-2007-036-S, Install Permanent Stay-Full Line for HPCI, Rev. 0
PB-2007-043-S, Torque Switch bypass for MO-3-23-016, Rev. 0
PB-2008-002-S, Installation of Emergency Service Water Hot Tap A Header, Rev. 0
PB-2008-004-S, Revision to Calculations PE-0166 Diesel Loading and PM-0123 EDG Fuel Oil
Consumption, Rev. 0
PB-2008-007-S, Change ,MCCB from 1M to TM, Rev. 0
PB-2008-008-S, Replacement of Obsolete SCRAM Discharge Volume (SDV) High Level
Switches LS-2-03-231 C/D, Rev. 0
PB-2008-013-S, Revise Thermal Power Uncertainty Calculations, Rev. 0
Attachment
A-2
PB-2008-022-S, Installation of Emergency Service Water Hot Tap B Header,
Rev. OPB-2008036-S, PB Units 2 and 3 Tech Spec. Bases Revision to B.3.3.2.1, Rev. 0
PB-2008-037-S, Temporary Power for Critical Loads from E324-0-A, Rev. 0
PB-2009-036-S, Reclassification of Stairwells 24 and 25 from Harsh Environment to Mild
Environment, Rev. 0
Modification Packages
ECR 96-04115, Evaluate Repair/Replacement for MO-89B Valve, Rev. 2
ECR 00-01131, Solid State Trip Devices for Load Center Breakers, Rev. 0
ECR 04-00352, Modify Orifice Plates Downstream of MO-3-1 0-089D, Rev. 0
ECR 05-00255, Mod to Provide Continuous Prelube for 'E1' EDG, Rev. 0
ECR 06-00023, Low Flow Thru HPSW Side of '3D' RHR HX During RT-0-032-300-3, Rev. 0
ECR 06-00099, 'E434' 480V LC Breakers - Replace OD Trip Devices with Solid State, Rev. 1
ECR 06-00461, Retire E-1 EDG Motor Driven Aux Pumps from Service, Rev. 2
ECR 07-00001, 'E424' 480V LC Breakers - Replace OD Trip Devices with Solid State, Rev. 0
ECR 07-00010, Install Permanent Stay-Full Line for HPCI, Rev. 1
ECR 07-00148, Lack of Spares for SCRAM Discharge Volume Level Switches, Rev. 0
ECR 07-00158, Peach Bottom 3, Cycle 16, Cycle Management Fuel Load, Rev. 1
ECR 07-00284, Replace 'E1' EDG Speed Switch with Older Model, Rev. 0
ECR 07-00342, Physical Mod Recommended for MO-3-23-016-0P for Margin, Rev. 0
ECR 07-00439, Change 1M MCCB to TM MCCB, Rev. 0
ECR 08-00011, Install Hot Tap to Support ESW System A Header Draining, Rev. 0
ECR 08-00017, Retire 'E3' EDG Aux Pumps and Associated Breaker Change to TM, Rev. 0
ECR 08-00033, 2BP098 Failing to Pump, Rev. 1
ECR 08-00074, Develop Core Spray Pump TS and 1ST Acceptance Criteria, Rev. 1
ECR 08-00104, Implementation of New Tech Specs for CR Envelope at PB, Rev. 0
ECR 08-00289, ESW Pipe Hot Tap B Header, Rev. 0
•
ECR 08-00371, Temporary Power for Critical Loads Fed from MCC E324-0-A, Rev. 1
ECR 09-00036, Peach Bottom 3 Soft Shutdown Sequence, Rev. 2
ECR 09-00339, 2BP098 DIW Equipment Drain Sump Pump No Flow, Rev. 0
ECR 09-00360, 'E1' EDG Tachometer Drive Coupling Shaft/Pin Replacement, Rev. 0
Calculations and Analysis
11187-M-22, PBOC/HELB Analysis Plant Model, Rev. 2
ME-0073, Core Spray Pump In-Service Testing Acceptance Criteria, Rev. 3
ME-0299, RCIC Pump Discharge Head Loss, Rev. 0
ME-0537, NPSH for HPCI and RCIC, Rev. 1
NEDC-32163P, Peach Bottom 2/3 SAFERIGESTR Analysis, Rev. 0
NEDC-32230P, Peach Bottom Power Rerate Project Engineering Report, dated March 1994
PE-0166, Emergency Diesel Generator Loading for Cases Defined by UFSAR, Rev. 8V, and 9
PE-0193, Coordination Study for 480VAC Load Centers and MCCs, Rev. 5M, 5P, and 6F
PE-0194, Coordination Study for 4kV 1E Switchgear, Rev. 3H
PM-0123, Diesel Generator Fuel Oil Consumption for 6 Days and 7 Days Operation with LOCA
Dependent Loads, Rev. 4
Attachment
A-3
PM-0957, Calculate HPCIIRCIC Room Temperature Profiles for 95°F River Temperature (Small
Break LOCA), Rev. 1
PM-1050, Heat BalancelThermal Power Uncertainty Analysis, Rev. OOC
PM-1 051, Heat BalancelThermal Power Uncertainty Analysis: Maintenance and Fail, Rev. 2
PM-1079, Design Analysis for High Pressure Service Water - Orifice Sizing, dated 3/9/07
PM-1081, Calculation of Maximum CRE Breach, Rev. 0
Corrective Action Reports
00227081
00328572
00370589
00388447
00478007
00580590
00600094
00600132
00601583
00606820
00618400
00640506
00656655
00662397
00723246
00727345
00792467
00852379
00935800
01040244*
01044407*
01044358*
* Issue report written as a result of inspection effort.
Drawings
31775, Order 6280-M245-1-3, 14 Inch Pressure Breakdown Restriction Orifices, dated 4/24/75
6280-A-487, Sht. 1, Barrier Plans, Rev. 11
6280-M-1-S, Electrical Schematic Diagram Rod Worth Minimizer System, Rev. 11
6280-M-315, Sht. 3, P&ID, Emergency Service Water and High Pressure Service Water
Systems, Rev. 53
6280-M-353, P&I Diagram Reactor Recirculation Pump System, Rev. 56
6280-M-384, Control Room HVAC, Rev. 40
6280-M-815, Sht. 3, Q.A.D., Emergency Service Water and High Pressure Service Water
Systems, Rev. 34
E-1617, Single Line Diagram E324 and E424 Emergency Load Centers, Rev. 64
E-1619, Single Line Meter and Relay Diagram, Rev. 32
E-1621, Single Line Diagram E324-T-B, E124-P-A, E124-0-A, and E34-0-A, Rev. 65
Surveillance and Modifications Acceptance Tests
M-055-005, 480V ITE Solid State Breaker Trip Device Testing, performed 7/15/08
M-C-700-231, K-Line Static Circuit Breaker Calibration, performed 5/18/08 and 8/21/08
RE-41 , InstaliationNerification of the 3D Monicore Thermal Operating Limits, performed
11/20109
RT-O-032-300-3, HPSW Pump, Valve and Flow Functional Test, performed 10/21/05 and
1/28/10
RT-O-052-251-2, 'E1' Diesel Generator Inspection PMT, performed 8/5/07 and 06128/09
RT-O-052-253-2, 'E3' Diesel Generator Inspection PMT, performed 5/13/09
RT-R-59C-500-3, NSSS Computer Calculation of Core Thermal Power, performed 10/19/09
ST-I-002-250-3, Core Flow Verification, performed 10/20109
ST-O-014-301-2, Core Spray Loop A Pump, Valve, Flow and Cooler Functional and In-Service
Test, performed 10/27109 and 1/25/10
Attachment
A-4
ST-0-014-301-3, Core Spray Loop A Pump, Valve, Flow and Cooler Functional and In-Service
Test, performed 1/6/10,4/14/09,7/15/09 and 10/22/09
ST-0-014-306-2, Core Spray Loop B Pump, Valve, Flow and Cooler' Functional and In-Service
Test, performed 2/11/09,5/16/09,11/18/09 and 2/11/10
ST-O-014-306-3, Core Spray Loop B Pump, Valve, Flow and Cooler Functional and In-Service
Test, performed 11/14/09 and 2/17/10
ST-0-032-301-3, HPSW Pump, Valve and Flow Functional and 1ST, performed 12/29/05
ST-0-054-754-2, 'E42' 4kV Bus Undervoltage Relays and LOCA LOOP Functional Test,
performed 9/30/08
ST-R-002-900-3, Reactivity Anomalies, performed 10/19/09
ST-R-002-910-3, Shutdown Margin Determination, performed 10/11/09
Procedures
AO 2A.16-2, Manual Adjustment of Recirculating Pump Seal Second Stage Pressure, Rev. 2
AO 33.6-0, ESW Pump Discharge Cross-Tie Operations, Rev. 1
CC-AA-103, Configuration Change Control for Permanent Physical Plant Changes, Rev. 20
CC-AA-309, Control of Design Analyses, Rev. 9
GP-3, Normal Plant Shutdown, Rev. 117
GP-30, Control Room Envelope Boundary Integrity, Rev. 1
LS-AA-104, Exelon 50.59 Review Process, Rev. 6
LS-AA-104-1000, 50.59 Resource Manual, Rev. 5
M-055-004, 480V K-Line Load Center Breaker Inspection, Rev. 6
MA-PB-724-005, 480V ABBIITE Load Center Breaker Maintenance, Rev. 1
ON-117, Loss of CRD Regulating Function, Rev. 15
SO 52A.8.A, Diesel Generator Daily Shutdown Inspection, Rev. 49
ST-0-62A-210-2, RWM Operability Check, Rev. 16
Work Orders
A1508236
A1558892
A1625842
A1637932
A1646537
C0214743
C0220918
C0221991
C0223475
C0223499
C0223502
C0226624
C0228635
C0229510
R0846490
R1087543
Vendor Manuals
385-A-VC-26, Low Voltage Air Magnetic Power Circuit Breakers, Rev. 2
385-A-VC-27, Power Shield Solid State Trip Device, Rev. 0
677-C-VC-32, LEFM Check Plus 2000 FC Flow Measurement System User Manual, dated
December 2002
L-200-VC-4 Limitorque Valve Operator Engineering Reference Manual, Rev. 0
NE-72-2, Instructions for HFD Circuit Breakers, Rev. 0
S-102-VC-2, Rod Worth Minimizer Detailed Design, Rev. 2
Attachment
A-5
Miscellaneous
6280-M1JJ-49-1, Core Spray Pump 2A, 2B, 2C, 2D, 3A, 3B, 3C, 3D Pump Curve, Rev. 1
982861-03, Focused Area Self Assessment, 2010 Mods and 50.59 Inspection, dated 2/1/10
ASME OM Code-2001
ER-463, Bounding Uncertainty Analysis for Thermal Power Determination at PB Unit 3 Using the
LEFM Check Plus System, Rev. 1
IAR TSTF-448, Control Room Envelope Habitability, dated 4/12/07
NEDO-32965-A, Reactor Stability Detect and Suppress Solutions Licensing Basis for Reload
Applications, dated August 1996
NEDO-33091-A, Improved BPWS Control Rod Insertion Process, dated July 2004
Purchase Order #90-031267, Rev. 0
Restricting Orifice Data Sheet for RO-3789 AlB/C/O, dated 2112/74
Restricting Orifice Data Sheet for RO-3800 AID, RO-3801 AID, dated 2/12/75
USAEC Safety Evaluation for Peach Bottom Atomic Power Station Units 2 and 3, dated 8/11/72
VTS-985-082881-01 R, Engine Prelube/Keepwarm System Requirements, dated 8/28/81
LIST OF ACRONYMS
AC
ADAMS
CFR
CRE
DRS
ECR
EDG
HPCI
HPSW
IR
1ST
LC
LOCA
LOOP
MCC
MCCB
MCREV
NEI
NRC
PARS
PMT
RHR
TS
UFSAR
Alternating Current
Agency-Wide Documents Access and Management System
Code of Federal Regulations
Control Room Envelope
Division of Reactor Safety
Engineering Change Request
Emergency Diesel Generator
High Pressure Coolant Injection
High Pressure Service Water
Issue Report
In-Service Test
Load Center
. Loss-of-Coolant-Accident
Loss-of-Offsite-Power
Motor Control Center
Molded Case Circuit Breaker
Main Control Room Emergency Ventilation
Nuclear Energy Institute
Nuclear Regulatory Commission
Publicly Available Records
Post Maintenance Test
Residual Heat Removal
Technical Specifications
Updated Final Safety Analysis Report
Attachment